The invention relates to nuclear reactors and more particularly to horizontal partitions of nuclear reactors that prevent thermally induced damage to internal structures of the reactor.
In designs of nuclear reactors that are well known in the art, the reactor includes a reactor vessel having an inlet and an outlet with a closure head disposed on the top thereof. A core disposed within the reactor vessel consists of an array of fuel assemblies containing nuclear fuel which produce heat in a commonly understood fashion. The fuel assemblies are supported at their lower end by a lower core plate and at their upper end by an upper core plate while a cylindrical metal member with an open top and bottom, known as a core barrel, surrounds the array of fuel assemblies. The outer surface of the core barrel together with the inner surface of the reactor vessel define an annular passage therebetween. The lower core plate is supported within the reactor core by a support structure attached to the inner wall of the reactor vessel in a manner to support the core within the reactor vessel. During reactor operation a coolant which in a fast breeder reactor may be liquid sodium is circulated through the reactor vessel passing in heat transfer relationship with the fuel assemblies thereby removing the heat produced by the fuel assemblies and carrying the heat to electrical generating equipment as is well known in the art.
In its circulation through the reactor vessel, the coolant flows into the reactor vessel through an inlet which may be located below the core near the bottom of the reactor vessel and enters an inlet plenum defined below the core. From the inlet plenum, the coolant flow separates into two paths; one passing through the core and the other passing through the annular passage around the core. The coolant following the path through the core enters the fuel assemblies near their lower end, passes in heat transfer relationship with the fuel assemblies thereby greatly raising the temperature of the coolant, and exits the fuel assemblies near their upper end flowing into an outlet plenum located above the core. In order to maintain the reactor vessel at a reasonable temperature, the other coolant path is traced through the annular passage around the core and into the outlet plenum where it mixes with the coolant that has passed through the core. This other coolant path is sometimes referred to as by-pass flow path. From the outlet plenum the coolant exits the reactor vessel through an outlet on the reactor vessel.
A problem associated with this arrangement occurs when the hot coolant that has passed through the core does not flow directly to the outlet on the reactor vessel, but rather flows non-uniformly from the outlet plenum of the core down the annular passage and comes into contact with the core support structure, the outer surface of the core barrel, and other related structures. The difficulty with this situation is that having been cooled by the by-pass flow of cooler coolant passing through the annular passage, the core support structure, the outer surface of the core barrel, and other related structures are at a much lower temperature than the coolant that has passed through the core. Therefore, the contact of the hot coolant with the cooler structures creates unacceptable thermal stresses in those structures. Furthermore, the contact of the hot coolant with those structures may not be continuous but may be cyclical which may lead to thermal fatigue of those structures. As a result it is necessary to prevent the hot coolant that has passed through the core from flowing in a reverse manner down the annular passage.
One method that attempts to prevent such a reverse flow is the device scheduled for use in the Fast Flux Test Facility (FFTF) which is under construction in the State of Washington. The FFTF device consists of an annular barrier supported on its underside and disposed in the annular passage between the core barrel and reactor vessel so as to substantially close the annular passage. However, the annular barrier is not sealed to the reactor vessel so that a reduced by-pass flow between the annular passage and the core outlet plenum, nevertheless, results. Flow experiments for FFTF indicate that flow instabilities and variable pressure distributions may exist in the outlet plenum which may induce oscillating flow patterns near the annular barrier. Since the annular barrier is not sealed to the reactor vessel, the variable flow patterns may cause, alternately, the flow of hot coolant and the cooler by-pass coolant across the edges of the annular barrier than are not sealed, resulting in high thermal stresses in the annular barrier that are not acceptable for common materials. Due to the size of the annular barrier and the severe environment in which it is placed, state of the art seals are not appropriate for use in preventing this alternate flow characteristic. Because state of the art seals may not be used, the FFTF device anticipates the use of Inconel as a suitable material capable of existing in such an environment, but Inconel is expensive and an undesirable material for such an application.
Another device that attempts to solve this problem is the one described in U.S. Pat. No. 4,001,079 filed Aug. 15, 1975, entitled "Thermal Baffle For Fast-Breeder Reactor", by J. A. Rylatt. The Rylatt patent describes a thermal baffle that includes an annular austenitic stainless steel baffle plate extending from the core barrel to the thermal liner of the reactor vessel. The baffle plate has on its lower surface inner and outer circumferential webs with radial webs extending between the circumferential webs and has on its upper surface insulation consisting of an austenitic stainless steel plate that attempts to maintain the baffle plate at a uniform temperature. At its inner support, the baffle plate is keyed to the core barrel while at its outer support the baffle plate rests on ledges of the thermal liner. In addition, flex seals are provided at each end in an attempt to minimize alternate flow thereacross. While the flex seals may prevent some alternate flow across the baffle plate on its lower side, the flex seals do not extend over the upper edge of the baffle plate leaving such upper edge exposed to alternate flow which thereby creates the possibility of thermal fatigue thereof.